1. Field of the Invention
The present invention relates to a fully passive decay heat removal system utilizing a partially immersed heat exchanger, the system comprising: a hot pool having received therein hot sodium heated by a nuclear reactor core; an intermediate heat exchanger which heat-exchanges with the sodium of the hot pool; a cold pool having received therein cold sodium cooled by passage through the intermediate heat exchanger, the cold pool being isolated from the hot pool; a support barrel extending vertically through the boundary between the hot pool and the cold pool, in which the upper end of the support barrel is higher than the liquid level of the hot pool, and the lower end thereof penetrates into the cold pool; a sodium-sodium decay heat exchanger received in the support barrel in order to remove decay heat from the inside of the nuclear reactor; a sodium-air heat exchanger which is provided at a position higher than the sodium-sodium decay heat exchanger; an intermediate sodium loop for heat removal which connects the sodium-sodium decay heat exchanger with the sodium-air heat exchanger; and a primary pump which pumps the sodium of the cold pool to the hot pool via the nuclear reactor core in a normal operating state to maintain the liquid level difference between the hot pool and the cold pool, such that the liquid level of the hot pool is higher than that of the cold pool, wherein a portion of the effective heat transfer tube of the sodium-sodium decay heat exchanger is immersed in the cold pool, particularly in a normal operating state, and the surface of the lower end of a shroud for the sodium-sodium decay heat exchanger, the lower end being immersed in the sodium of the cold pool, has perforated holes.
2. Description of the Prior Art
Sodium-cooled fast reactors which are currently developed are provided with a safety grade decay heat removal system to remove decay heat produced in the reactor core due to emergency reactor shutdown when the normal heat removal paths, which are connected to the reactor core, an intermediate heat exchanger (IHX) and a steam generator (SG), are lost.
Till now, in a variety of liquid-metal-cooled fast reactors (hereinafter referred to as “liquid-metal reactors”), including sodium-cooled fast reactors, a passive safety-grade decay heat removal system has been used to enhance safety. For example, a decay heat removal system for a pool-type liquid-metal reactor is designed such that the heat of the system can be effectively removed by natural coolant circulation using the thermal inertia of a hot pool disposed above the reactor core outlet.
In the case of a conventional large-scale pool-type nuclear reactor, as shown in FIG. 1, a fully immersed-type decay heat exchanger (DHX) 3 through which the sodium of an intermediate sodium loop 2 for heat removal flows is installed in a hot pool 1 filled with the sodium of the primary circuit, and a sodium-air heat exchanger (AHX) 4 is installed at the upper portion of the nuclear reactor building. Thus, through a density difference resulting from a height difference of more than 20 m formed between the source to which heat is transferred (i.e., DHX) and the source from which heat is removed (i.e., AHX), sodium for heat removal naturally circulates in the intermediate sodium loop 2 for heat removal, which is separately provided, and the heat of the primary circuit is removed by the air which is the ultimate heat sink. This method is called “direct reactor cooling” (DRC).
However, in the design concept of such direct reactor cooling, the heat exchange between the hot pool 1 and the sodium-air heat exchanger occurs continuously even during normal operation. Thus, in order to prevent heat loss during normal operation, the flow rate of air into the sodium-air heat exchanger 4 is controlled by installing a separate isolation valve 5 in the intermediate sodium loop 2 or installing a damper 6 in an inlet/outlet pipe for the air, so that solidification of the intermediate sodium loop 2 for heat removal is prevented and the amount of heat loss during normal operation is controlled. Accordingly, although the intermediate sodium loop 2 for heat removal performs the heat removal function by natural coolant circulation, a starting signal for operating the system is produced either by the intervention of an operator or by a trip signal, and thus the system is designed such that active devices provided with a deriving unit such as the damper 6 or the isolation valve 5 are operated. For this reason, strictly speaking, the system shown in FIG. 1 is barely classifiable as a fully passive system.
To overcome the incompleteness of this passive system, another prior art shown in FIG. 2 utilizes a fully passive decay heat removal system for a pool-type liquid-metal reactor. In such a fully passive decay heat removal system, a sodium-sodium decay heat exchanger (DHX) support barrel 12 made of a vertical cylindrical tube, which communicates a hot pool 10 with a cold pool 11, is disposed in the reactor pool area, and a sodium-sodium decay heat exchanger 13 is disposed on the free surface of the cold pool using the liquid level difference between the hot pool 10 and the cold pool, which is maintained by the head of a primary pump (not shown), such that it does not make direct contact with the sodium.
Thus, the system shown in FIG. 2 is designed such that direct contact between the sodium-sodium decay heat exchanger 13 and the sodium of the cold pool 11 during normal operation can be fundamentally prevented, so that unnecessary heat loss during normal operation can be minimized without requiring the use of either an isolation valve in an intermediate sodium loop 15 for heat removal or a damper in the air inlet/outlet of a sodium-air heat exchanger 14.
Particularly, such a design concept adopts the concept in which the decay heat of the system is dispersed into the ultimate heat sink (atmosphere) by a vigorous convection heat transfer between the sodium of the primary loop and the intermediate sodium loop 15 for heat removal, only when the sodium liquid level rises due to a shutdown of the nuclear reactor and the primary pump resulting from loss of the normal heat removal paths. Thus, this design concept makes it possible to realize a passive decay heat removal system for large-scale nuclear reactors, which adopts a fully passive concept and can improve operational reliability.
However, the prior art shown in FIG. 2 does not utilize an air flow control device such as a damper in the sodium-air heat exchanger 14 to realize the fully passive concept. Thus, during the winter season when the temperature of outdoor air falls below zero or during the refueling operation when the temperature of the nuclear reactor pool sodium falls to 200° C. (for reference, the average temperature of the sodium of the nuclear reactor pool is 467° C.), there is a possibility for the sodium of the intermediate sodium loop to be solidified or frozen. The melting point of sodium is about 98° C., and thus when an outdoor air temperature of −40° C. which is adopted as the design standard is introduced into the air inlet of the sodium-air heat exchanger 14, there is a high possibility for the sodium in the sodium heat transfer tube of the sodium-air heat exchanger 14 to be solidified. In this prior art, if solidification occurs in the intermediate sodium loop 15, particularly the sodium heat transfer tube of the sodium-air heat exchanger 14, the paths for removing decay heat by natural circulation from the nuclear reactor pool into the atmosphere will be closed, so that the safety of the nuclear power plant will be seriously deteriorated. Accordingly, a method capable of preventing the sodium of the intermediate sodium loop from solidifying during the entire operating period of a nuclear power plant, including the operational standby period, is necessarily required.
The solidification of sodium in the intermediate sodium loop of the prior passive decay heat removal circuit (PDRC) adopting the fully passive concept and direct reactor cooling occurs for the following reason. The circuit is designed such that heat transfer from the sodium pool of the primary circuit to the sodium of the intermediate sodium loop during normal operation is achieved only by the radiation heat transfer mechanism without direct contact of the sodium, and thus the temperature of sodium in the intermediate sodium loop does not sufficiently rise depending on outdoor air conditions due to the influence of radiation heat transfer resistance which is much higher than convection or conduction heat transfer resistance.
Thus, in order to satisfy the design standard, the supply of heat should be stably secured, such that the solidification of sodium in the intermediate sodium loop is prevented during the entire operating period of the nuclear power plant, including the normal operating mode. However, there are disadvantages in that it is difficult only by radiation heat transfer to supply heat in an amount sufficient for preventing the solidification of sodium in the intermediate sodium loop in the normal operating mode and for forming the fundamental circulation of a coolant in the operational standby mode and in that the operational reliability of the safety system is significantly reduced, because design uncertainty in interpretative methodologies for the surface emissivity and radiation heat transfer of the internal structures of the nuclear reactor is great. Particularly, the prior art shown in FIG. 2 provides neither a method nor equipment, which copes with the case in which the temperature of outdoor air falls below zero, and serves to prevent sodium in the intermediate sodium loop from solidifying at the refueling temperature (200° C.) of a nuclear power plant.
Furthermore, the most urgent problem to be solved in the prior art can be considered to be the operating performance of the fully passive decay heat removal circuit (PDRC) in the case in which the pump of the primary circuit is operated without being shut down after the shutdown of the nuclear reactor. In this case, because the liquid level difference between the hot pool and the cold pool can be maintained by the operation of the pump, the liquid level of the cold pool can rise. For this reason, it is impossible to perform the smooth heat removal function of the fully passive decay heat removal circuit (PDRC), and a serious failure to perform the natural function of the safety system can occur. Accordingly, there is needed a design concept in which a given quantity of heat removal function can be provided even when the liquid level difference between the hot pool and the cold pool is maintained. However, the prior art does not describe a proper method for ensuring operational reliability and stable performance related therewith.
In addition, in the transition stage in which the normal heat removal function is lost so that the safety system is regularly operated, there is a possibility in which an overflow of sodium introduced from the hot pool can be formed into undesirable bypass flow in a ring-shaped space between the sodium-sodium decay heat exchanger (DHX) support barrel and the sodium-sodium decay heat exchanger (DHX) shroud without being introduced into the sodium-sodium decay heat exchanger in the DHX support barrel. In this case, it is expected that the flow of operating fluid which is used for heat removal will decrease, leading to deterioration in the heat removal performance of the sodium-sodium decay heat exchanger. However, the prior art provides neither any method nor improved design concept for a loop structure which can also reduce pressure loss without deteriorating the formation of the primary circulation flow passing through the sodium-sodium decay heat during the transition stage and normal operation.